Similar Documents at Ginna |
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Category:CORRESPONDENCE-LETTERS
MONTHYEARML17250B3041999-10-20020 October 1999 Forwards Changes to Tech Specs Bases for Re Ginna Nuclear Power Plant.Change Bars Indicate Those Revs Which Have Been Incorporated IR 05000244/19990081999-10-14014 October 1999 Forwards Insp Rept 50-244/99-08 on 990809-0919.Severity Level IV Violation of NRC Requirements Occurred & Being Treated as non-cited Violation Consistent with App C of Enforcement Policy ML17265A7651999-10-0808 October 1999 Forwards Fifteen Relief Requests That Will Be Utilized for Ginna NPP Fourth Interval ISI Program That Will Start on Jan 1,2000.Attachment 1 Includes Summaries & Detailed Description of Each Relief Request ML17265A7631999-10-0505 October 1999 Requests Approval for Use of Relief Request Number 43 to Address Volumetric Examination Limitations (Less than 90%) Associated with a & B RHR Heat Exchanger Outlet Nozzle to Shell Welds.Approval Is Requested by Dec 31,2000 ML17265A7641999-10-0505 October 1999 Requests Approval for Use of Relief Request Number 42 to Address Volumetric Examinations Limitations (Less than 90%) Associated with Eight Class 1 Identified Welds or Areas of Reactor Pressure Vessel ML20212J3561999-09-30030 September 1999 Forwards Four Copies of Re Ginna NPP Training & Qualification Plan for Security Officers, Rev 7,dtd 990930. Synopsis of Changes,Encl.Encl Withheld Per 10CFR73.21 ML20212J3801999-09-30030 September 1999 Forwards Four Copies of Rev to Re Ginna NPP Security Plan.Rev Changes Contingency Weapons Available to Response Force to Those Most Effective in Current Defensive Strategy.Encl Withheld Per 10CFR73.21 IR 05000244/19990051999-09-24024 September 1999 Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs ML17265A7571999-09-24024 September 1999 Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs IR 05000244/19992011999-09-24024 September 1999 Forwards Insp Rept 50-244/99-201 (Operational Safeguards Response Evaluation) on 990621-24.No Violations Noted. Primary Purpose of Osre to Assess Licensee Ability to Respond to External Threat.Insp Rept Withheld ML17265A7461999-08-31031 August 1999 Submits Response to NRC Administrative Ltr 95-03,rev 2, Availability of Reactor Vessel Integrity Database,Version, Dtd 990726 ML17265A7401999-08-26026 August 1999 Requests Approval for Use of Relief Request Number 35 Re Use of ASME Section XI Code,1995 Edition,1996 Addenda.Code Will Be Used to Develop Plant Fourth 10-year Interval ISI Program on Class 1,2 & 3 Components ML17265A7411999-08-26026 August 1999 Forwards LER 99-004-01 Re Plant Being Outside Design Basis Due to Containment Recirculation Fan Moisture Separator Vanes Being Incorrectly Installed.Part 21 Notification of 990512 Is Being Rescinded ML17265A7451999-08-23023 August 1999 Forwards fitness-for-duty Performance Data Rept for Six Months Ending 990630,per 10CFR26.71(d) ML17250B3021999-08-23023 August 1999 Informs That Util & NRC Had Conference Call on 990816 to Review Approach in Responding to Questions,As Result of Questions Re Main Steam Check Valve Performance Included in Insp Rept 50-244/99-05,dtd 990806 ML17265A7271999-07-30030 July 1999 Forwards 10CFR21 Interim Rept Per Reporting of Defects & Noncompliance,Section 21 (a) (2).Interim Rept Prepared Because Evaluation Cannot Be Completed within 60 Days from Discovery of Deviation or Failure to Comply ML17265A7151999-07-23023 July 1999 Forwards LER 99-007-01 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Reactor Trip.Further Investigation of Event Identified Addl Corrective Actions ML17265A7061999-07-22022 July 1999 Forwards LER 98-003-02,re Actuations of CR Emergency Air Treatment Sys.All Creats Actuations,Including Those Originally Believed to Be Valid Actuations,Were,In Fact Invalid Actuations ML17265A7191999-07-21021 July 1999 Forwards Ginna Station ISI Rept for Refueling Outage Conducted in 1999 ML17265A7141999-07-21021 July 1999 Withdraws Relief Request 35 for Plant Inservice Insp Program Section XI Requirements,Submitted on 980806.Licensee Plans to Resubmit Relief Request,Which Includes Addl Level of Detail,In Near Future ML17265A7041999-07-16016 July 1999 Submits Info Re Specific Licensing Actions Which May Be Expected to Generate Complex Reviews,In Response to Administrative Ltr 99-02,dtd 990603 ML17265A6911999-06-30030 June 1999 Responds to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants.Gl 98-01 Requested Response on Status of Facility Y2K Readiness by 990701.Readiness Disclosure Attached ML17265A6871999-06-22022 June 1999 Forwards Response to RAI Made During 990225 Telcon Re GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves. Calculation Encl. Encl ML17265A6841999-06-21021 June 1999 Informs That Util Wishes to Amend Extend of Alternate Exams Provided for Relief Request Re ISI Program ASME Section XI Require Exams for First 10-Yr Interval for Containment ML17265A6741999-06-15015 June 1999 Submits Annual ECCS Rept IAW 10CFR50.46(a)(3)(ii) Requirements.No Changes Have Been Made to Large Break LOCA PCT & Small Break LOCA PCT ML17265A6751999-06-11011 June 1999 Responds to NRC RAI Re Licensee GL 96-05 Program.Encl Info Verifies That Util Is Implementing Provisions of JOG Program on MOV Periodic Verification ML17309A6551999-06-0707 June 1999 Responds to NRC 990310 RAI Re Verification of Seismic Adequacy of Mechanical & Electrical Equipment ML17265A6671999-06-0101 June 1999 Requests Approval of Ginna QA Program for Radioactive Matl Packages,Form 311,approval Number 0019.Ginna QA Program for Station Operation, Was Most Recently Submitted to NRC by Ltr Dtd 981221 & Supplemented on 990301 ML17265A6641999-05-25025 May 1999 Forwards Addl Info on Use of GIP Method a for Re Ginna Nuclear Power Plant.Copy of Re Ginna Station USI A-46 Outlier Resolution Table as Requested ML17265A6611999-05-24024 May 1999 Forwards LER 99-007-00 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Rt.Util Is Planning to Submit Suppl to LER by 990730 ML20206U0921999-05-13013 May 1999 Forwards Four Copies of Rev R to Gnpp Security Plan,Per Provisions of 10CFR50.54(p).Rev Clarifies Armed Response Team Assignments & Does Not Degrade Physical Security Effectiveness.Rev Withheld,Per 10CFR73.21 ML17265A6461999-05-12012 May 1999 Forwards 1998 Annual Radioactive Effluent Release Rept & 1998 Annual Radiological Environ Operating Rept, for Re Ginna NPP ML17265A6411999-05-12012 May 1999 Forwards LER 99-004-00 IAW 10CFR50.73 & 10CFR21.Further Assessment Will Be Provided in Suppl to LER by 990630 ML20206E7221999-04-29029 April 1999 Forwards Four Copies of Rev Q to Re Ginna Nuclear Power Plant Security Plan,Per 10CFR50.54(p).Changes Do Not Degrade Physical Security Effectiveness.Encl Withheld,Per 10CFR73.21 ML20206H6911999-04-22022 April 1999 Forwards Info Requested During Informal Telcon on 980408 Concerning Upcoming Osre at Ginna Station.Info Requested Listed.Without Encls ML17265A6271999-04-19019 April 1999 Forwards Rev 0 & Rev 1 to Colr,Cycle 28 for Re Ginna NPP, Per TS 5.6.5 ML17309A6501999-04-14014 April 1999 Forwards Revised Ginna Station EOPs & Procedures Index ML17265A6121999-03-29029 March 1999 Forwards Rept on Status of Decommissioning Funding for Re Ginna Npp,For Which Rg&E Is Sole Owner,Per 10CFR50.75. Data Presented Herein,Current as of 981231 ML17265A6041999-03-24024 March 1999 Forwards LER 99-001-00,IAW 10CFR50.73(a)(2)(ii)(B) & 10CFR21.Addl Analyses Are Being Performed to Support Future Cycle Operation & Supplemental LER Is Scheduled to Be Submitted by 990618 ML17265A5671999-03-0101 March 1999 Forwards Application for Amend to License DPR-18,to Revise TSs Battery Cell Parameters Limit for Specific Gravity (SR 3.8.6.3 & SR 3.8.6.6).Supporting Tss,Encl ML17265A5641999-03-0101 March 1999 Forwards Response to NRC 990217 RAI Concerning Changes to QA Program for Re Ginna Station Operation.Rg&E Is Modifying Changes Requested in 981221 Submittal.Modified QA Program,Encl ML17265A5551999-02-25025 February 1999 Informs That Util Is in Process of Revising fitness-for-duty Program,Developed in Accordance with 10CFR26.Util Will Continue to Use Dept of Health & Human Svcs Certified Test Facility for Majority of Tests During Yr ML17265A5561999-02-22022 February 1999 Forwards FFD Performance Data Rept for Six Months Ending 981231,per 10CFR26.71(d) ML17265A5451999-02-12012 February 1999 Forwards Simulator Four Year Certification Rept,Per 10CFR55.45(b)(5)(ii) ML17309A6491999-02-12012 February 1999 Forwards Ginna Station EOPs ML17265A5431999-02-0909 February 1999 Supplements 980806 Relief Request with Attached Table.Util Third 10-Yr ISI of Reactor Vessel Being Performed During 1999 Refueling Outage,Beginning on 990301 ML17265A5361999-02-0202 February 1999 Forwards Response to NRC 981203 RAI Re Resolution of Unresolved Safety Issue USI A-46.Util Does Not Agree with NRCs Interpretation.Detailed Bases,Encl ML17311A0691999-01-25025 January 1999 Forwards Revs to Ginna Station Emergency Plan Implementing Procedures (Epips).Previous Rev Had Incorrect Effective Date ML17311A0671999-01-14014 January 1999 Forwards Revised Emergency Operating Procedures for Re Ginna NPP ML17265A5131999-01-12012 January 1999 Forwards Revised Cover Page for Ginna Station Technical Requirements Manual (Trm), Rev 7,correcting Error in Effective Date 1999-09-30
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML17250B3041999-10-20020 October 1999 Forwards Changes to Tech Specs Bases for Re Ginna Nuclear Power Plant.Change Bars Indicate Those Revs Which Have Been Incorporated ML17265A7651999-10-0808 October 1999 Forwards Fifteen Relief Requests That Will Be Utilized for Ginna NPP Fourth Interval ISI Program That Will Start on Jan 1,2000.Attachment 1 Includes Summaries & Detailed Description of Each Relief Request ML17265A7631999-10-0505 October 1999 Requests Approval for Use of Relief Request Number 43 to Address Volumetric Examination Limitations (Less than 90%) Associated with a & B RHR Heat Exchanger Outlet Nozzle to Shell Welds.Approval Is Requested by Dec 31,2000 ML17265A7641999-10-0505 October 1999 Requests Approval for Use of Relief Request Number 42 to Address Volumetric Examinations Limitations (Less than 90%) Associated with Eight Class 1 Identified Welds or Areas of Reactor Pressure Vessel ML20212J3561999-09-30030 September 1999 Forwards Four Copies of Re Ginna NPP Training & Qualification Plan for Security Officers, Rev 7,dtd 990930. Synopsis of Changes,Encl.Encl Withheld Per 10CFR73.21 ML20212J3801999-09-30030 September 1999 Forwards Four Copies of Rev to Re Ginna NPP Security Plan.Rev Changes Contingency Weapons Available to Response Force to Those Most Effective in Current Defensive Strategy.Encl Withheld Per 10CFR73.21 IR 05000244/19990051999-09-24024 September 1999 Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs ML17265A7571999-09-24024 September 1999 Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs ML17265A7461999-08-31031 August 1999 Submits Response to NRC Administrative Ltr 95-03,rev 2, Availability of Reactor Vessel Integrity Database,Version, Dtd 990726 ML17265A7401999-08-26026 August 1999 Requests Approval for Use of Relief Request Number 35 Re Use of ASME Section XI Code,1995 Edition,1996 Addenda.Code Will Be Used to Develop Plant Fourth 10-year Interval ISI Program on Class 1,2 & 3 Components ML17265A7411999-08-26026 August 1999 Forwards LER 99-004-01 Re Plant Being Outside Design Basis Due to Containment Recirculation Fan Moisture Separator Vanes Being Incorrectly Installed.Part 21 Notification of 990512 Is Being Rescinded ML17250B3021999-08-23023 August 1999 Informs That Util & NRC Had Conference Call on 990816 to Review Approach in Responding to Questions,As Result of Questions Re Main Steam Check Valve Performance Included in Insp Rept 50-244/99-05,dtd 990806 ML17265A7451999-08-23023 August 1999 Forwards fitness-for-duty Performance Data Rept for Six Months Ending 990630,per 10CFR26.71(d) ML17265A7271999-07-30030 July 1999 Forwards 10CFR21 Interim Rept Per Reporting of Defects & Noncompliance,Section 21 (a) (2).Interim Rept Prepared Because Evaluation Cannot Be Completed within 60 Days from Discovery of Deviation or Failure to Comply ML17265A7151999-07-23023 July 1999 Forwards LER 99-007-01 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Reactor Trip.Further Investigation of Event Identified Addl Corrective Actions ML17265A7061999-07-22022 July 1999 Forwards LER 98-003-02,re Actuations of CR Emergency Air Treatment Sys.All Creats Actuations,Including Those Originally Believed to Be Valid Actuations,Were,In Fact Invalid Actuations ML17265A7191999-07-21021 July 1999 Forwards Ginna Station ISI Rept for Refueling Outage Conducted in 1999 ML17265A7141999-07-21021 July 1999 Withdraws Relief Request 35 for Plant Inservice Insp Program Section XI Requirements,Submitted on 980806.Licensee Plans to Resubmit Relief Request,Which Includes Addl Level of Detail,In Near Future ML17265A7041999-07-16016 July 1999 Submits Info Re Specific Licensing Actions Which May Be Expected to Generate Complex Reviews,In Response to Administrative Ltr 99-02,dtd 990603 ML17265A6911999-06-30030 June 1999 Responds to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants.Gl 98-01 Requested Response on Status of Facility Y2K Readiness by 990701.Readiness Disclosure Attached ML17265A6871999-06-22022 June 1999 Forwards Response to RAI Made During 990225 Telcon Re GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves. Calculation Encl. Encl ML17265A6841999-06-21021 June 1999 Informs That Util Wishes to Amend Extend of Alternate Exams Provided for Relief Request Re ISI Program ASME Section XI Require Exams for First 10-Yr Interval for Containment ML17265A6741999-06-15015 June 1999 Submits Annual ECCS Rept IAW 10CFR50.46(a)(3)(ii) Requirements.No Changes Have Been Made to Large Break LOCA PCT & Small Break LOCA PCT ML17265A6751999-06-11011 June 1999 Responds to NRC RAI Re Licensee GL 96-05 Program.Encl Info Verifies That Util Is Implementing Provisions of JOG Program on MOV Periodic Verification ML17309A6551999-06-0707 June 1999 Responds to NRC 990310 RAI Re Verification of Seismic Adequacy of Mechanical & Electrical Equipment ML17265A6671999-06-0101 June 1999 Requests Approval of Ginna QA Program for Radioactive Matl Packages,Form 311,approval Number 0019.Ginna QA Program for Station Operation, Was Most Recently Submitted to NRC by Ltr Dtd 981221 & Supplemented on 990301 ML17265A6641999-05-25025 May 1999 Forwards Addl Info on Use of GIP Method a for Re Ginna Nuclear Power Plant.Copy of Re Ginna Station USI A-46 Outlier Resolution Table as Requested ML17265A6611999-05-24024 May 1999 Forwards LER 99-007-00 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Rt.Util Is Planning to Submit Suppl to LER by 990730 ML20206U0921999-05-13013 May 1999 Forwards Four Copies of Rev R to Gnpp Security Plan,Per Provisions of 10CFR50.54(p).Rev Clarifies Armed Response Team Assignments & Does Not Degrade Physical Security Effectiveness.Rev Withheld,Per 10CFR73.21 ML17265A6461999-05-12012 May 1999 Forwards 1998 Annual Radioactive Effluent Release Rept & 1998 Annual Radiological Environ Operating Rept, for Re Ginna NPP ML17265A6411999-05-12012 May 1999 Forwards LER 99-004-00 IAW 10CFR50.73 & 10CFR21.Further Assessment Will Be Provided in Suppl to LER by 990630 ML20206E7221999-04-29029 April 1999 Forwards Four Copies of Rev Q to Re Ginna Nuclear Power Plant Security Plan,Per 10CFR50.54(p).Changes Do Not Degrade Physical Security Effectiveness.Encl Withheld,Per 10CFR73.21 ML20206H6911999-04-22022 April 1999 Forwards Info Requested During Informal Telcon on 980408 Concerning Upcoming Osre at Ginna Station.Info Requested Listed.Without Encls ML17265A6271999-04-19019 April 1999 Forwards Rev 0 & Rev 1 to Colr,Cycle 28 for Re Ginna NPP, Per TS 5.6.5 ML17309A6501999-04-14014 April 1999 Forwards Revised Ginna Station EOPs & Procedures Index ML17265A6121999-03-29029 March 1999 Forwards Rept on Status of Decommissioning Funding for Re Ginna Npp,For Which Rg&E Is Sole Owner,Per 10CFR50.75. Data Presented Herein,Current as of 981231 ML17265A6041999-03-24024 March 1999 Forwards LER 99-001-00,IAW 10CFR50.73(a)(2)(ii)(B) & 10CFR21.Addl Analyses Are Being Performed to Support Future Cycle Operation & Supplemental LER Is Scheduled to Be Submitted by 990618 ML17265A5641999-03-0101 March 1999 Forwards Response to NRC 990217 RAI Concerning Changes to QA Program for Re Ginna Station Operation.Rg&E Is Modifying Changes Requested in 981221 Submittal.Modified QA Program,Encl ML17265A5671999-03-0101 March 1999 Forwards Application for Amend to License DPR-18,to Revise TSs Battery Cell Parameters Limit for Specific Gravity (SR 3.8.6.3 & SR 3.8.6.6).Supporting Tss,Encl ML17265A5551999-02-25025 February 1999 Informs That Util Is in Process of Revising fitness-for-duty Program,Developed in Accordance with 10CFR26.Util Will Continue to Use Dept of Health & Human Svcs Certified Test Facility for Majority of Tests During Yr ML17265A5561999-02-22022 February 1999 Forwards FFD Performance Data Rept for Six Months Ending 981231,per 10CFR26.71(d) ML17265A5451999-02-12012 February 1999 Forwards Simulator Four Year Certification Rept,Per 10CFR55.45(b)(5)(ii) ML17309A6491999-02-12012 February 1999 Forwards Ginna Station EOPs ML17265A5431999-02-0909 February 1999 Supplements 980806 Relief Request with Attached Table.Util Third 10-Yr ISI of Reactor Vessel Being Performed During 1999 Refueling Outage,Beginning on 990301 ML17265A5361999-02-0202 February 1999 Forwards Response to NRC 981203 RAI Re Resolution of Unresolved Safety Issue USI A-46.Util Does Not Agree with NRCs Interpretation.Detailed Bases,Encl ML17311A0691999-01-25025 January 1999 Forwards Revs to Ginna Station Emergency Plan Implementing Procedures (Epips).Previous Rev Had Incorrect Effective Date ML17311A0671999-01-14014 January 1999 Forwards Revised Emergency Operating Procedures for Re Ginna NPP ML17265A5131999-01-12012 January 1999 Forwards Revised Cover Page for Ginna Station Technical Requirements Manual (Trm), Rev 7,correcting Error in Effective Date ML17265A5111999-01-11011 January 1999 Requests Relief Per 10CFR50.55a(a)(3)(ii) from Certain Requirements of Section XI of ASME Bp&V Code for ISI Program.Relief Requests 37,38 & 39 Encl ML17265A5101999-01-11011 January 1999 Requests Relief Per to 10CFR50.55a(a)(3)(ii) from Certain Requirements of Section XI of ASME B&PV Code for ISI Program.Relief Request 40 Encl 1999-09-30
[Table view] Category:UTILITY TO NRC
MONTHYEARML20059L4481990-09-18018 September 1990 Forwards Responses to Rer 900709-13 Team Visit Findings, Per .Responses Withheld (Ref 10CFR73.21) ML17309A4491990-09-13013 September 1990 Forwards Slides Presented by Facility & Westinghouse to NRC in 900724 Meeting.Encl Withheld ML17262A1331990-09-11011 September 1990 Responds to Violations & Several Unresolved Items Noted in SSFI Rept 50-244/89-81.Update of Appropriate Unresolved Items Encl.Specific Actions Re All NRC Unresolved Items Being Tracked to Completion ML17261B1511990-08-29029 August 1990 Forwards Semiannual Radioactive Effluent Release Rept Jan- June 1990 & Rev 3 to Process Control Program for Ginna Station, Per Tech Specs 6.9.1.4 & 6.16,respectively ML17261B1481990-08-28028 August 1990 Lists Understanding of Issues Util Planning to Address Re Containment Integrity,Per 900718 Telcon.Any Concerns or Action Items Different from Listed Submittal Should Be Provided to Util Prior to NRC 900905 & 06 Visit to Plant IR 05000244/19880261990-08-20020 August 1990 Submits Update on Util Plans to Provide Appropriate Ventilation to Intermediate Bldg Clean Side,Per Violation Noted in Insp Rept 50-244/88-26.Extensive Program to Manually Close All External Openings Implemented ML17261B1501990-08-20020 August 1990 Submits Update on Util Plans to Provide Appropriate Ventilation to Intermediate Bldg Clean Side,Per Violation Noted in Insp Rept 50-244/88-26.Extensive Program to Manually Close All External Openings Implemented ML17261B1371990-08-17017 August 1990 Forwards Rev 0 to, Inservice Insp Rept for Third Interval (1990-1999),First Period,First Outage (1990) at Re Ginna Nuclear Power Plant. ML17309A4481990-08-16016 August 1990 Responds to NRC 900717 Ltr Re Violations Noted in Insp Rept 50-244/90-80.Corrective Actions:Westinghouse Drawing E-2508 Approved & Issued to Central Records ML17261B1351990-08-15015 August 1990 Provides Info & Assessment on Integrity of Connection of Containment Bldg to Foundation,Per 900808 Telcon.Util Believes That Existing Condition Not Safety Concern & No Reason Exists to Suspect Joint Will Not Perform Function ML17261B1361990-08-14014 August 1990 Responds to Commitment Tracking Concerns Noted in Insp Rept 50-244/90-09 & Planned Corrective Actions.Util Confirms Commitments Dates for Implementation of Effective Shelf Life Program & Comprehensive Preventive Maint Program for Parts ML17261B1331990-08-13013 August 1990 Recommits to Performing Enhanced primary-to-secondary Leak Rate Monitoring,Per NRC Bulletin 88-002 ML17261B1261990-07-30030 July 1990 Clarifies Commitment Made in 900316 RO Re Restoring Inoperable Fire Damper I-411-21-P.Util Plans to Design Removable Track Which Will Allow Charcoal Drawer to Be Manipulated.Definitive Schedule Will Be Provided in 60 Days ML20055H9291990-07-23023 July 1990 Forwards Revised Page 6:4 of Plant Security Plan.Page Withheld Per 10CFR73.21 ML17261B1091990-07-20020 July 1990 Advises That Structural Evaluations of Containment Sys Being Performed in Response to Containment Integrity Insp ML17309A4471990-07-17017 July 1990 Forwards Decommissioning Rept, for Plant,Per 10CFR50.33(k) & 50.75(b) ML17261B0881990-07-13013 July 1990 Provides Update to Util 860616 Ltr Re Implementation of NUREG-0737,Item 6.2,Suppl 1, Emergency Response Capability. ML17261B0921990-07-11011 July 1990 Responds to 900709 Request for Addl Info Re Containment Integrity Insp.Util Will Provide Preliminary Results by 900716 Re Where Groundwater Entering Annular Access Area. Meeting Proposed for Wk of 900723 ML20044B0481990-07-10010 July 1990 Discusses Review of Station Blackout Documentation,Per 10CFR50.63.Licensee Will Complete Enhancements to Station Blackout Documentation Identified in Attachment 1 as Indicated & Other Items Will Be Completed within 2 Yrs ML17309A4461990-07-0909 July 1990 Responds to Generic Ltr 90-04 Re Status of Licensee Implementation of Generic Safety Issues Resolved W/ Imposition of Requirements or Corrective Actions ML17261B0931990-07-0909 July 1990 Responds to Generic Ltr 88-14, Instrument Air Supply Sys Problems Affecting Safety-Related Equipment. Util Completed All Actions Requested in Generic Ltr & Will Retain Documentation Verification for Min of 2 Yrs ML17261B0901990-07-0909 July 1990 Advises That Licensee Will Install Containment Isolation Signal Going to Valve AOV 745 by End of 1992 Refueling Outage,Per Util 900608 Ltr Notifying of Condition Outside Design Basis of Plant Under 10CFR50.72 Criterion ML17250B2151990-06-29029 June 1990 Forwards Application for Amend to License DPR-18, Reformatting Auxiliary Electrical Sys Tech Specs & Action Statements for Offsite & Onsite Power Sources Available for Plant Auxiliaries ML17261B0851990-06-28028 June 1990 Forwards LER 89-016-02 Re 891117 Failure of Safety Injection Block/Unblock Switch Which Could Render Both Trains of Safety Injection Sys Inoperable.Also Reported Per Part 21 ML17250B1991990-06-0808 June 1990 Responds to NRC 900509 Ltr Re Violations Noted in Insp Rept 50-244/89-81.Corrective Actions:Improved Battery Load Profile Developed Incorporating Calculational Improvements Contained in Current Industry Std IEEE 485-1983 ML17250B1951990-06-0505 June 1990 Responds to Generic Ltr 89-08, Erosion/Corrosion Induced Pipe Wall Thinning. Util Developed Erosion/Corrosion Program for Single & Two Phase Sys Consistent W/Requirements of NUREG-1344 & NUMARC 870611 Rept ML17250B1851990-06-0101 June 1990 Forwards Application for Amend to License DPR-18,providing Guidance for Action Statements Associated W/Power Distribution Limit Specs ML17261B0691990-06-0101 June 1990 Discusses Testing Frequency for Insp of Incore Neutron Monitoring Sys Thimble Tubes,Per NRC Bulletin 88-009.Thimble Tube Indicating Greatest Wear Recently Repositioned in Effort to Minimize Future Wear ML17250B1811990-05-31031 May 1990 Forwards Addl Info Re Response to Notice of Violation from Insp Rept 50-244/90-04.Info Withheld (Ref 10CFR73.21) ML17250B1801990-05-29029 May 1990 Responds to NRC Bulletin 90-001, Loss of Fill-Oil in Transmitters Mfg by Rosemount. Util Does Not Need to Develop Enhanced Surveillance Program to Monitor Currently Installed Transmitter Based Upon Limited Installed Quantity ML17250B1841990-05-29029 May 1990 Responds to NRC Bulletin 89-002, Stress Corrosion Cracking of High Hardness Type 410 Stainless Steel (SS) Internal Pre-Load Bolting.... Valves Disassembled & 410 SS Studs Removed & Visually & Liquid Penetrant Examined ML17261B0961990-05-29029 May 1990 Responds to NRC 900427 Ltr Re Violations Noted in Insp Rept 50-244/90-07.Corrective Actions:Worker Records Immediately Corrected Indicating That Worker Received No Significant Exposure for Time While Error Occurred ML17250B1721990-05-18018 May 1990 Forwards Rev 1 to Summary Exam Rept for 1990 Steam Generator Eddy Current Insp. Rept Summarizes Observations & Corrective Actions Resulting from Insps Performed During 1990 Outage ML20042G8931990-05-0808 May 1990 Forwards Revised Emergency Operating Procedures,Including Rev 3 to AP-RHR.2,Rev 1 to ES-0.3 & Rev 11 to ES-1.3 ML17261B0611990-04-26026 April 1990 Advises of Changes to 890201 Commitments Made Under Programmed Enhancement Response to Generic Ltr 88-17.Changes Will Not Reduce Capability to Operate Safely in Reduced Inventory Condition or Scope of Programmed Enhancements ML17261B0591990-04-17017 April 1990 Advises That No Interlocks Required for RHR motor-operated Valves 701 & 720 Based on Present Arrangement.Listed Failures Would Have to Occur in Order for Potential Overpressurization of RHR Sys to Occur ML17261B0461990-04-12012 April 1990 Responds to Issues Discussed During 900323 Telcon Re Inservice Testing Program Status & Relief Request.Current Test Methodology for Seat Leakage Acceptable Based on Application of Direct Measurement Sys ML17262A1431990-04-12012 April 1990 Responds to NRC 900202 Ltr Re Weaknesses Noted in Insp Rept 50-244/89-80.Corrective Actions:Procedure ES-0.3 Modified to Provide Guidance for Rapid Cooldown & Depressurization W/ & W/O Reactor & Vessel Instrumentation Sys ML17250B1451990-04-0606 April 1990 Discusses Impact of SER Issuance for Inservice Insp & Inservice Testing Programs & Advises That Timing Will Not Affect Util Implementation Plans for Programs,Except as Listed ML17250B1441990-04-0505 April 1990 Responds to NRC 900309 Ltr Re Violations Noted in Insp Rept 50-244/90-02.Corrective Actions:Positive Indicator Will Be Installed for All Check Valves to Enhance Visible Positive Position Verification Ability & to Avoid Confusion ML17261B0401990-03-30030 March 1990 Advises That Final Results of Station Blackout Documentation Review Will Be Submitted to NRC on or About 900501 ML17261B0321990-03-28028 March 1990 Forwards Annual Rept of ECCS Model Revs as Applicable to Facility,Per 10CFR50.46.Mods to Model for Small Break LOCA Do Not Affect Calculated Peak Clad Temp ML17261B0331990-03-28028 March 1990 Forwards Info Re Reactor Vessel Issues,Per 900305 Telcon Concerning Change of License Expiration Date ML17261B0231990-03-26026 March 1990 Responds to NRC 890222 Ltr Re Violations Noted in Insp Rept 50-244/89-17.Corrective Actions:Personnel Verified Safety Injection Block/Unblock Switch in Proper Position & Operator Procedure 0-1.1 Changed as Indicated ML17261B0281990-03-23023 March 1990 Forwards Addl Info Re Proposed Tech Spec Amend Concerning Use of Reconstituted Fuel,Per 900315 Telcon.During Fuel Assembly Reconstitution,Failed Fuel Rods Will Be Placed W/ Filler Rods ML17261B0221990-03-22022 March 1990 Provides Revised Test Schedule for motor-operated Valve Diagnostic Test Program,Per IE Bulletin 85-003.NRC Notification of Changes to Valve Operability Program Required by Generic Ltr 89-10,dtd 890828 ML17261B0411990-03-20020 March 1990 Forwards Summary of Onsite Property Damage Coverage Currently in Force at Plant,Per 10CFR50.54(w)(4) ML17261B0151990-03-19019 March 1990 Responds to Generic Ltr 89-19, Safety Implication of Control Sys in LWR Nuclear Plants (USI A-47). Overfill Protection Provided Through Trip Bistables in Reactor Protection Racks Powered from 120-volt Instrument Buses ML17261B0101990-03-15015 March 1990 Forwards Application for Amend to License DPR-18,allowing Use of Reconstituted Fuel Assemblies ML20012D2831990-03-14014 March 1990 Responds to Issues Discussed During 900307 Telcon W/Util & Eg&G Re Inservice Testing Program Status & Relief Request. Valves 5960A & 5960B Will Be Disassembled to Verify Forward Flow.Relief Requests PR-8 & VR-25 Encl 1990-09-18
[Table view] |
Text
REGULATORY I ORMATION DISTRIBUTION SYST " (RIDS)
ACCESSION NBR;8506040273 DOC ~ DATE: 85/05/24 NOTARIZED: NO DOCKET' FACIL'::50-244 Robert Emmet Ginna Nuclear Plant~ Unit lg Rochester G 05000244
'AUTH,NAME AUTHOR AFFILIATION KOBERiR ~ WE Rochester Gas L Electric Corp, RECIP,NAME RECIPIENT AFFILIATION ZWOLINSKIpA ~ Operating Reactors Branch 5
SUBJECT:
Responds to 850221 request for addi info to complete safety evaluation re performance testing of safety ~ relief valvesp per NUREG"0737iltem II.D, 1.Feedline break analysis performed in 197$ using conservative assumptions, DISTRIBUTION CODE! A046D COPIES RECEIVED LTR TITLE: OR Submittal: THI Action Plan Rgmt NUREG"0737 8 NUREG 0660
/ 'NCL ~: SIZE+ Z<
NOTES:NRR/DL/SEP icy, 05000244 OL $ 09/19/69 RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME'RA LTTR ENCL" ID CODE/NAME LTTR ENCL ORB5 BC 01 7 7 INTERNAL: ACRS 34 10- 10 ADM/LFMB 1 0 ELD/HDS4 1- 0 IE/DEPER DIR 33 1 IE/DEPER/EPB 3 3 NRR PAULSONeW ~ 1 1 NRR/DHFS DEPY29 l NRR/DL DIR 1 1 NRR/DL/ORAB 18 3 NRR/DSI/ADRS 27 1 1 NRR/DSI/AEB 1 NRR/DSI/ASB 1 1 B 1 1 NRR/DST DIR 30 1 1 G ILE 04 1 1 RGNl 1 1 EXTERNAL; 24X 1 1 LPDR 03 1 NRC PDR 1 1 NSIC 05 1 1 NOTES!
TOTAL NUMBER OF COPIES REQUIRED; LTTR 41 ENCL 39
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I /IIIII/ I /IIIIIIIIII/ j 55A5t'LECTRIC ROCHESTER GAS AND CORPORATION o 89 EAST AVENUE, ROCHESTER, N.Y. 14649.0001 ROGER W. KOBER Vee mESIOENT TCLSPHONC ELECTRIC & STEAM PAOOVCTION ARc* eood 7ld 546.2700 May 24< 1985 Director of Nuclear Reactor Regulation Attention: Mr. John A. Zwolinski< Chief Operating Reactors Branch No. 5 U.S. Nuclear Regulatory Commission Washington'.C. 20555
Subject:
NUREG-0737i Item II.D.l Performance of Testing of Relief and Safety Valves R. E. Ginna Nuclear Power Plant Docket No. 50-244
Dear Mr. Zwolinski:
Your letter dated February 21i 1985 requested that RG6E provide additional information in order that the NRC Staff can complete a Safety Evaluation of the performance testing of safety and relief valves. A response to each of the Staff requests is contained in Attachment A.
V r truly y urs<
R er N./
W. Kober 85 bpy ~~3 85p5 DOCK Ospp0244 p
PDR
Attachment A Response to NRC Request for Additional Information Dated February 21, 1985 NUREG-0737 Item II.D.l The Westinghouse valve inlet fluid conditions report stated that liquid discharge through both the safety and Power Operated Relief Valves (PORVs) is predicted for a FSAR feedline break event. The Westinghouse report gave expected peak pressure and pressurization rates for some plants having a FSAR feedline break analysis. The Robert E. Ginna Unit 1 plant was not included in this list of plants having such a FSAR analysis. Nor does the R. E. Ginna plant specific submittal address the FSAR feedline break event. NUREG-0737, however, requires analysis of accidents and occurrences referenced in Regulatory Guide 1.70, Revision 2, and one of the accidents so required is the feedline break.
Provide a discussion on the feedwater line break event. either justifying that it does not apply to this plant or identifying the fluid pressure and pressurization rate, fluid temperature, valve flow rate, and time duration for the event. Assure that the fluid conditions were enveloped in the EPRI tests and that the time period of water relief in the EPRI test was as long as expected at the plant.. Demonstrate operability of the safety valves and PORV's for this event and assure that the feedline break event was considered in analysis of the piping system.
Page 1 Nay 24, 1985
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RESPONSE
A feedline break analysis was performed in 1973 using very conservative assumptions in order to maximize the time delay to reactor trip and minimize steam generator heat removal capability. Additional conservatism was provided in the analysis by making the assumption that no auxiliary feedwater was available for 10 minutes in order to demonstrate that a manually initiated standby auxiliary feedwater system was sufficient, to remove decay heat following postulated accidents. The results were found to be acceptable (see SEP Topic XV-6, NRC SER of 9-04-81).
Although the full diameter feedline break downstream of the feedwater check valve was assumed along with a 10 minute auxiliary feedwater delay for purposes of analysis, such a combination is not required by the Ginna design basis. No full diameter feedwater line break is postulated which can disable all three automatically initiated auxiliary feedwater pumps. Thus this analysis is not J
appropriate to be used",for the purposes of pressurizer relief and safety valve system qualification. Nuclear plants such as Ginna were licensed prior to issuance of Regulatory Guide 1.70, Revision 1, and were not required to consider the feedline break as part of their design basis.
Page 2 Nay 24, 198S
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- In valve operability discussions on cold overpressurization transients, the submittal only identifies conditions for water discharge transients. According to the Westinghouse valve inlet fluid conditions report, however, the PORVs are expected to operate over a range of steam, steam-water, and water conditions because of the potential presence of a steam bubble in the pressurizer.
To assure that the PORVs operate for all cold overpressure events, II discuss the range of fluid conditions for expected types of fluid discharge and identify the test data that demonstrate operability for these cases. Since no low pressure steam tests were performed for the relief valves, confirm that the high pressure steam tests demonstrate operability for the low pressure steam case for both opening and closing of the relief valves.
RESPONSE
With regard to the cold overpressure (COP) event, the maximum temperature and pressure conditions that can be achieved at the PORV inlet coincidentally occur for steam bubble operation.
Since pressure is normally maintained below the PORV setpoint, the maximum steam and saturated liquid pressure maintained in the pressurizer during startup and shutdown operations in anticipation of the COP event would occur at the PORV setpoint. This pressure is 435 psig, and the corresponding temperature is 456'F. To allow for the calculated potential overshoot of the 435 psig setpoint, the limiting .pressure is 535 psig.
Page 3 May 24, 1985
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Accordingly, the potential worst, case scenarios (i.e., maximum temperature and pressure conditions) for PORV discharge during a COP event would be:
- 1. Discharge of saturated steam at P < 535 psig and T
< 456'F (steam in upper phase of pressurizer).
- 2. Discharge of saturated water at P 5 535 psig and T
< 456'F (saturated water in pressurizer).
- 3. Discharge of subcooled water at, P < 535 and T S 456'F (some mixing of colder RCS water with saturated pres-surizer water).
- 4. Scenario 1 followed by Scenario 2
- 5. Scenario 2 followed by Scenario 3
- 6. Scenario 1 followed by Scenario 2 followed by Scenario 3.
EPRI test conditions for the PORV's were chosen based on expected inlet fluid conditions. Tests were limited but, designed to confirm operability over a full range of expected inlet conditions.
Steam, steam-to-water and water flow tests were conducted.
Results of these tests can be found in EPRI report EPRI NP-2670-LD, Volume 8. Although steam tests were conducted only at the higher pressures, that satisfactory operation would also result at the less severe lower pressures. This can be confirmed by the high pressure versus low pressure water tests where successful valve operation was observed.
Page 4 Nay 24, 1985
Results from the EPRI tests on the Crosby safety valve indicate that. the test blowdowns exceeded the design value of 5 percent for both "as installed" and "lowered" ring settings. If the blowdowns expected for the plant (see Question 4) also exceeded 5 percent, the higher blowdowns could cause a rise in pressurizer water level such that water may reach the safety valve inlet line and result in a steam-water flow situation., Also the pressure might be sufficiently decreased such that adequate cooling might, not be achieved for decay heat removal. Discuss these consequences of higher blowdowns I if increased blowdowns are expected.
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RESPONSE
The impact on plant safety of pressurizer relief valves blowdowns in excess of 5 percent for R. E. Ginna was evaluated. The results of this evaluation showed no adverse effects on plant safety.
Relief valve blowdowns in excess of that assumed in the R. E.
Ginna Reload Transition Safety Report (RTSR) (submitted by letter dated December 20, 1983 from John E. Naier, RGSE, to Harold R. Denton USNRC) will have the following effects on the events in which relief valve actuation occurs:
- 1) Increased pressurizer water level during and following the valve blowdown,
- 2) Iower pressurizer pressure during and following valve blowdown,
- 3) Increased inventory loss through the relief valve.
Page 5 May 24, 1985
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The impact of the increased relief valve blowdowns with respect to the above effects was evaluated for the two R. E. Ginna RTSR events in which relief valve actuation occurs, (i.e., Loss of Load and Locked Rotor).
For the Loss of Load event, results from sensitivity analyses performed by Westinghouse for 4 loop plants were used for the evaluation. Similar results are expected for 2 loop plants.
These analyses investigated the effects of different blowdown rates on the Loss of Load event. The results of these analyses showed only marginal increases in pressurizer water volume and the maximum pressurizer water levels were well below the level at which liquid relief would occur. Peak RCS pressures were shown to be unaffected by the increased blowdowns. The increased blowdowns did result in lower pressurizer pressure and increased RCS inventory loss, however, these had no adverse impact on the event and adequate decay heat removal was maintained.
For the Locked Rotor event, increased relief valve blowdowns have little impact on the event. As analyzed and presented in the R.E. Ginna RTSR, the opening and closing of the relief valve occurs over a short time period ( 3 seconds). As a result, there is little change in either pressurizer level or RCS inventory.
Increased relief valve blowdowns would have no impact. on peak pressure, peak clad temperature, or DNBR as these occur prior to the closing of the relief valve.
Page 6 Nay 24, 1985
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The submittal states that an evaluation of safety valve ring settings is being performed by RGSE and Westinghouse to verify that the ring settings used will result in proper valve performance on a plant specific basis. Identify the final ring settings selected as a result of this evaluation. Since EPRI tests on the Crosby 3K6 and 6M6 safety valves were used to assess performance of the 4K26 valve of R. E. Ginna, identify which EPRI tests on the 3K6 and 6M6 valves had ring settings representative of those
'sed on the plant 4K26 valve. Identify the expected blowdowns corresponding to the plant ring settings and explain how these blowdowns were extrapolated or calculated from test data. Verify that with the ring settings used the valves can perform their pressure relief function and the plant can be safely shutdown with the blowdown, back pressures, and fluid conditions occurring at the plant.
RESPONSE
Valve ring settings for the Ginna safety valves were developed using Crosby production test methods and will have performance characteristics similar to those valves tested at EPRI with "as-shipped" ring settings. Details of these tests were discussed in the original submittal and can be found in EPRI reports EPRI NP-2770-LD, Volumes 5 and 6. Blowdowns measured on the Ginna valves at Crosby during the production testing for each valve were equal to or less than 5 percent.
Page 7 May 24, 1985
UESTION 5 Results from EPRI tests on the Crosby 3K6 and 6M6 safety valves were used to evaluate performance of the Crosby 4K26 valve of R. E. Ginna. The EPRI test results indicate that steam flow rates in excess of rated flow were achieved. A flow rate deter-mination for the R. E. Ginna valves, however, depends on the y - ~
specific ring settings used at the plant.. Thus, provide a,demon-stration that the plant safety valves will pass their rated flow at the ring settings used.
RESPONSE
As noted in Table 4.4 of EPRI Report NP-2770-LD, Volume 6, the Crosby 6M6 test valve achieved rated flow for each of the tests reported at 3 percent, accumulation regardless of the ring setting used in the test. A review of EPRI Tables 4-3 and 4-4 in Volume 5 of EPRI Report NP-2770-LD reveals that, for steam tests of the 3K6 valve where blowdown was measured to be less than 10 percent, flow rates of 119-122 percent of rated flow at 3 percent accumu-lation were reported. The EPRI tables indicate that lower than rated flows occurred at blowdowns greater than 15 percent.
Crosby standard production tests indicate 5 percent blowdown with the "as-shipped" ring settings. These are the ring settings currently used on the installed valves at, Ginna Station. This is within the range of both the 3K6 and 6M6 tests where rated flow was achieved; therefore, rated flow can be expected for the 4K26 R. E. Ginna valves.
Page 8 May 24, 1985
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UESTION 7 NUREG-0737, Item II.D.1 requires that the plant-specific PORV control circuitry be qualified for design-basis transients and accidents. Please provide information which demonstrates that this re'quirement has been fulfilled.
RESPONSE
NUREG-0737 Item II.D.l required that licensees conduct testing to quality the reactor coolant system relief and safety valves under w expected operating conditions for design basis transients and accidents. Licensees were to determine the expected valve operating conditions through the use of analyses of accidents and anticipated operational occurrences. The single failures applied to these analyses-were to be chosen so that the dynamic forces on the safety and relief valves were maximized. Test pressures were to be the highest predicted by conventional safety analysis procedures.
The required analyses were performed and the expected valve operating conditions were given in the RGE report submitted Narch 4, 1983. The limiting conditions were developed based upon FSAR, extended high pressure injection and cold overpressurization events. The limiting events for two loop plants, including Ginna, were loss of load and locked rotor transients.
Page ll Nay 24, 1985
No relief or safety valve discharge will occur for extended high pressure injection events. Cold overpressurization events will not cause adverse environmental conditions in containment. The limiting pressure event, a locked rotor, is analyzed in Section 15.3 of the Ginna UFSAR. The transient is terminated by safety valve operation discharging at 20 ft /sec over a period of approxi-mately four seconds. The discharge of the safety and relief valves is to the pressurizer relief tank which is designed to ll I lt 1'
condense and cool a di'scharge equivalent to 110 percent of the",
full power pressurizer steam space (UFSAR Section 5.4.8) or approximately 400 ft . Thus", no adverse'nvironment, is expected to occur as a result of the most limiting transient requiring operation of the relief or safety valves. Those transients that do cause an adverse environment, such as ZOCAs, steam line breaks, feedwater line breaks or other high energy line breaks do not require operation of the PORVs to terminate the transient or mitigate the consequences of the event. Therefore, the Ginna power operated relief valves (PORVs) are capable of fulfilling their design function.
In response to 10 CFR 50.49, RG&E provided, in a submittal dated August, 30, 1984, a discussion of the selection process used to define electrical equipment required to operate in a harsh environment.
Based upon that methodology, which was found acceptable in the NRC's SER'ated February 28, 1985, RGSE has determined that the pressurizer PORVs and control circuitry are not required to be qualified to operate in a harsh environment.
Page 12 May 24, 1985
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UESTION 8 Bending moments are induced on the safety valves and PORVs during the time they are required to operate bacause of discharge loads and thermal expansion of the pressurizer tank and inlet piping.
Nake a comparison between the predicted plant moments with the moments applied to the tested valves to demonstrate that the operability of the valves will not be impaired.
RESPONSE
The bending moments induced on the safety and relief valves tested by EPRI exceeded the bending moments predicted for the Ginna safety and relief valves. The maximum moment tested for the 6M6 valve was during test 908 and was 298.75 in-K. The largest moment predicted for the safety valves for Ginna is 73.27 in-K, thus demonstrating functionability for the Ginna safety valves.
1 h
I F r if Likewise, a bending moment, of 43.0 in-K was induced at. the inlet of the Copes-Vulcan PORV test valve per EPRI test 64-CV,-174-2S.
The largest bending moment, predicted for the Ginna relief valves is 30.48 in-K, thus demonstrating functionability for the Ginna relief valves.
Page 13 Nay 24, 1985
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The submittal does not address the NUREG-0737 II.D.l requirement that the operability of the PORV block valves be demonstrated. A test program on block valves was performed, though, which is described in the EPRI/Marshall Electric Motor Operated Valve Interim Test Data Report. In this test program the Limitorque SMB-000-5 motor operator that is used at the R. E. Ginna plant was not tested.. Since the SMB-000-5 operator is smaller than any tested, explain how these tests results or other test data can be used to demonstrate operability'f the motor operator.'"
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RESPONSE
V Three differences exist between the R'. E. Ginna block valves and those (Velan) block valves tested by EPRI. First the motor operator installed on the R. E. Ginna block valves is a Limitorque SMB-000-5 versus the Limitorque SB-00-15 operator used with the EPRI test valve. The speed of the two valves is different, 40 seconds for the R. E. Ginna valve and 10 seconds for the EPRI valve and finally the seat bore of the EPRI valve (2.625 inches) is larger than that of the R. E. Ginna (2.25 inches) valve. The R. E. Ginna block valve operators provide an output thrust at the as-shipped torque switch setting that exceeds that required to close by approximately 26 percent.
Page 14 May 24, 1985
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E Since the valves at R. E. Ginna are of similar design except for the motor operators (the motor operators are different. due to the different closing time requirements) to the ones tested by EPRI and the output thrust is greater than required, the valves at R. E. Ginna will exhibit performance equal to or better than the EPRI test valves. The fact that the seat bore is smaller on the Ginna block valve results in a smaller thrust required to close the valve as compared to the test valve at the same differential pressure.
Page 15 Nay 24, 1985
UESTION 10 The submittal indicates that a discharge piping backpressure of 350 psig is developed. Describe how this value for backpressure was determined. Among the EPRI tests performed on the Crosby 3K6 and 6M6 safety valves, thre was only one test performed with a hot loop seal and high backpressure. This was a loop seal-steam-to-water transition test on the 3K6 valve, which resulted in valve chatter. Explain how the results of cool loop seal tests with high backpressure can be used to show that the safety valves of the Ginna,plant"can successfully dischar'ge hot loop Jt t seal water followed by steam under high plant backpressures.
RESPONSE
The maximum safety valve discharge piping backpressure during and following discharge of the safety valves calculated from our thermal hydraulic analysis is less than 357 psia. The thermal hydraulic analysis was described in the submittal dated March 4, 1983 from John E. Maier, RG&E to Dennis M. Crutchfield, USNRC.
The EPRI tests showed that the Crosby safety valves always lifted on steam after loop seal water discharge whether it be hot or cold. Dynamic backpressures measured during the EPRI tests were approximately 9 to 29 percent of the setpoint; however, Crosby valves employ a balanced bellows design that is relatively insensitive to backpressure effects as noted by EPRI on page S-5 of EPRI report NP-2770-LD, Volume 5. Backpressure effects and loop seal water temperature, therefore, should not impact valve functionability.
Page 16 May 24, 1985
UESTION 11 In the analysis of the hot, loop seal discharge, case, which is representative of the R.E. Ginna installation, the loop seal temperature distribution was assumed to be consistent with the distribution of EPRI test 917. A misrepresentation of the temperatures in the loop seal could have a significant effect on the calculated forces acting on the piping system. Therefore, provide verification that the temperatures assumed for the loop seal are accurate.
RESPONSE
RG8E does not currently have data verifying that the temperatures in the loop seal are consistent with those used in the safety valve discharge piping analysis. Therefore, in order to provide'ssurance that the temperatures in the loop seal are representative of those assumed in the piping analysis, further work has been initiated. During the 1985 refueling outage at Ginna physical measurements in the loop seal area were taken to permit the PI I'I I
performance of a detailed thermal" i/
analysis,and',designof an I4 improved insulation system for the safety valve loop seals. This new insulation system will ensure that the temperatures achieved are consistent with those assumed in the previous piping analysis.
Installation of this modification is presently scheduled for the 1986 refueling outage.
Page 17 Nay 24, 1985
UESTION 12 According to results of EFRI tests, high frequency pressure oscillations of 170-260 Hz typically occur in the piping upstream of the safety valve while loop seal water passes through the valve. The submittal refers to an evaluation of this phenomenon that is documented in the Westinghouse report WCAP 10105 and states that the acoustic pressures occurring prior to and during safety valve discharge are below the maximum permissible pressure.
The study discussed in the Westinghouse report determined the maximum permissible pressure for the inlet piping and established the maximum allowable bending moment for Level C Service Conditions E
in the inlet piping based on the maximum transient pressure measured or calculated. While the internal pressures are lower than the maximum permissible pressure, the pressure oscillations could potentially excite high frequency vibration modes in the piping, creating bending moments in the inlet piping that should be combined with moments from other appropriate mechanical loads.
Provide one of the following: (a) a comparison of the allowable bending moments established in WCAP 10105 for Level C Service Conditions with the bending moments induced in the plant piping by dynamic motion and other mechanical loads or (b) justification for other alternate allowable bending moments with a similar comparison with moments induced in the plant piping.
RESPONSE
The R.E. Ginna piping system response including the safety valve loop seal region is due to frequencies less than 100 HZ. The Page 18 May 24, 1985
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frequency of the forces a'nd moments in'the 170-260 HZ range
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potentially induced by the pressure oscillations is significantly greater than this frequency. The upper limit of significant content for similar systems is also much less than 'this (170-260 HZ) range. Industry data indicates that only frequencies of 100 HZ or less 'are meaningful. The EPRI data confirms this. Consequently, no significant bending moment during the pressure oscillation phase of the transient will occur.
In the submittal, pressure stresses based upon a design pressure of 2580 psig were included with the bending moments resulting from the deadweight and the safety valve discharge piping loads.
Because of the time phasing of the pressure oscillation (during water slug discharge through the safety valve) and the discharge piping loads (subsequent to water slug discharge through the valve) this pressure term and moment term were not added. They do not occur coincidentally. A comparison of the intensified bending moments from the stress evaluation and the allowable moment presented in WCAP-10105 shows that, all values are below the allowables. Specifically, the maximum allowable moment from Table 4-7 of WCAP-10105 for 4-inch schedule 160 piping for an internal pressure of 5000 psi is 176 in-kips. The moments for the sum of deadweight and water slug discharge for the components listed in Table 6-14 of the submittal at nodes 6110, 5010, and 6120 respectively, are 47.2, 52.3, and 47.3 in-kips.
Page 19 Nay 24, 1985
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